This present invention relates generally to radiation shielding materials, radiation shielding containers and methods for preparing the same. More particularly, the present invention relates to radiation shielding materials incorporating uranium dioxide and/or uranium carbide and containers for radioactive materials incorporating these shielding materials. This invention also relates to methods for preparing uranium dioxide and uranium carbide microspheres for use in the radiation shielding materials of the present invention.
Storage, transportation, and disposal of radioactive waste, such as spent nuclear fuel (xe2x80x9cSNFxe2x80x9d), high level waste (xe2x80x9cHLWxe2x80x9d), mixed waste, and low level radiation waste is a growing problem in the United States and abroad. In 1995, the Department of Energy (DOE) estimated that the commercial SNF inventory was about 30,000 metric tonnes initial heavy metal (xe2x80x9cMTIHMxe2x80x9d) and is expected to exceed 80,000 MTIHM within two decades. (1 tonnes=1 metric ton=2,205 pounds). Adding DOE""s own inventory of SNF and HLW raises the domestic total to nearly 90,000 MTIHM.
Unfortunately, it appears that many U.S. commercial nuclear power plants do not have sufficient existing storage capacity to accommodate future SNF discharges. Moreover, much of the DOE""s SNF and HLW inventory is currently located in unlicensed storage structures. Many of these storage structures will have to be upgraded or replaced, and the SNF and HLW relocated. Thus, there is a need for improved radiation shielding materials and radiation shielding containers incorporating these shielding materials for the storage, transportation, and disposal of radioactive materials, including, in particular, SNF waste.
Two principal types of storage methods are generally used for SNF: wet and dry. In wet storage, the SNF is typically immersed in a lined, water-filled pool which performs the dual functions of shielding and heat removal with the assistance of and reliance on active systems. Wet storage of SNF is generally required for a given period of time (about 5 years) after the SNF has been discharged from a nuclear reactor. Thereafter, the SNF can be placed into long term dry storage. Dry storage encompasses a wide spectrum of structures that house the fuel in a dry inert gas environment, with an emphasis on passive system design and operation. In dry storage, the radioactive material is typically disposed in dry vaults or dry casks. Dry vault installations generally utilize a concrete building or other concrete structure for radiation shielding. Dry cask storage, on the other hand, utilizes prefabricated containers including an appropriate shielding material. Because dry cask storage is usually accomplished more quickly and cheaply, it is generally preferred over vault storage. Dry cask storage is also preferred at sites having an existing infrastructure for receipt, examination, and loading of SNF for economic and scheduling reasons.
The design and manufacture of a suitable container for the dry storage of SNF involves a variety of factors, such as (1) subcriticality assurance, (2) shielding effectiveness, (3) structural integrity (i.e., containment), (4) thermal performance, (5) ease of use, (6) cost, and (7) environmental impact. Other factors that may affect the selection process are whether the design has been previously licensed and actually used to store SNF, or, if the design has not been licensed, its perceived ability to meet applicable regulations and standards.
The first factor in designing a storage container is the maintenance of subcriticality. In dry storage, the subcriticality design relies on controlling the fissile SNF and SNF spacing, and sometimes incorporates the use of neutron-absorbing materials. The subcriticality control design of dry storage containers is generally acceptable and does not typically provide any discriminating factors for selecting one design over another.
The second factor in designing a storage container is shielding effectiveness. Shielding effectiveness affects both onsite worker and public dose rates during the loading and subsequent storage of SNF. Both neutron and gamma ray shielding must be provided and ensured throughout the life of the storage system. Dry storage technology relies on a number of solid shielding materials, sometimes in combination, to reduce gamma and neutron dose rates. The most common solid shielding materials are different forms of concrete (low-density, high-density, or hydrogenated), metal (ductile cast iron, carbon steel, stainless steel, lead), borated resin, and polyethylene (for neutrons). Often, in order to function effectively, metal shielding materials must be combined with additional materials to enhance their neutron absorbing ability.
The third factor in designing a storage container is structural integrity (i.e., containment). Structural integrity ensures that the confinement boundary around the SNF is maintained under all operational and postulated accident conditions. All SNF storage technologies are required to meet the same standards for structural integrity in accordance with appropriate codes. Therefore, the selection of a suitable storage technology will include consideration of the structural integrity of the proposed design.
The fourth factor in container design is thermal performance. With the exception of steel and cast iron, most shielding materials have inherent limiting temperatures (i.e., a maximum allowable temperature that is lower than the fuel cladding temperature limit). Shielding material thermal limits include both absolute values of temperature and, in the case of concrete, temperature gradients that create thermal stresses. Adequate decay heat removal is vital to preventing degradation of the fuel cladding barrier to fission product releases.
Dry storage containers rely on a combination of conduction, convection (natural or forced), and radiation heat transfer mechanisms to maintain fuel cladding temperatures below appropriate long term storage limits. In particular, metal casks rely on a totally passive system for heat removal. The fuel decay heat, in an encapsulating inert gas atmosphere canister, is transferred to the canister""s walls by a combination of radiation and conduction heat transfer. The canister walls, which are in contact with the metal cask wall, transfer this heat by conduction. At the outside of the metal cask, the heat is removed by conduction and natural convention to the environment. Metal cask typically are not susceptible to thermal limits, since the metals have a higher temperature limit than that of the fuel cladding. However, in those embodiments where the metal casks incorporate additional neutron shielding materials their favorable heat-transfer properties may be compromised.
As with metal casks, concrete casks use a passive heat removal system. Concrete casks, however, have an inherent vulnerability, because concrete""s thermal conductivity is a factor of 10 to 40 lower than that of metal. Thus, in order to remove fuel decay heat and stay below both the fuel cladding and concrete temperature limits, concrete casks must include labyrinthine airflow passages that allow natural convection-driven air to enter the cavity enclosing the canister inside the concrete and then exit through higher elevation passages in the concrete to the environment. The need for these airflow passages introduces the possibility of an accident in which adequate heat removal is reduced or eliminated because of inlets and/or outlets that are blocked by debris, snow, or even nests and hives. As a result,concrete casks require surveillance of their air inlet and outlet flow passages, thereby increasing the associated life-cycle costs and personnel radiation exposures.
The fifth factor in designing a storage container is ease of use, which is defined as the lack of complexity involved in the operation and maintenance of SNF. As noted above, the existence of labrynthine air passages in concrete casks means that additional operation and maintenance is required. Ease of use, however, is also related to the complexity associated with loading, transport, and storage of SNF. Thus, the weight and size of containers are also of particular importance. For example, since many existing storage sites are already equipped with a crane in the storage and receiving facility, it is desirable to utilize containers with weights that are within the typical crane capacity of 45 to 91 tonnes. Metal casks generally cannot be used with such cranes, because the weight of a fully-shielded metal cask loaded with a large number of SNF elements can easily exceed the 91 tonnes limit. Thus, even though metal casks have desirable heat transfer characteristics, the additional weight and size associated with metal systems limits their applicability.
Additional size and weight limits are imposed when containers are transported. The U.S. Department of Transportation and state highway regulations generally limit the gross weight of a waste-carrying road vehicle to about 80,000 pounds. Since the typical tractor trailer weighs about 30,000 pounds, the weight of a transportation container and its contents should not exceed about 50,000 pounds. Heavier weights can be transported by rail, but maximum container widths (diameters) are limited to approximately 9 feet to allow for adequate clearance between tracks. U.S. Nuclear Regulatory Commission regulations require that the container provide certain levels of shielding and be capable of sustaining certain impact stresses without yielding the waste. The end result of these regulations is that much of the available weight for the transportation container and its contents must be expended in providing adequate shielding and a shell that can withstand the designated impact stresses. The resulting thickness of the container walls leaves a relatively small amount of space in the container for SNF.
The sixth factor in designing a storage container is cost. Concrete casks are generally the least expensive, with a typical cost of about $350,000 to $550,000, versus $1 million to $1.5 million for their metal cask counterparts.
The seventh factor in designing a storage container is environmental impact. Over time, environmental mechanisms can degrade storage containers, possibly exposing the SNF directly to groundwater or air. Storage containers and shielding materials that minimize degradation are preferred for long term storage and disposal.
In summary, metal casks are desirable because they are known to provide effective heat transfer and structural integrity. Unfortunately, metal casks are heavier and more expensive than concrete casks. Furthermore, in most SNF applications, metal casks must incorporate separate neutron shields, which may compromise their favorable heat transfer properties.
Thus, there is a significant need for improved, lower weight and higher heat-transfer shielding materials and, also, for containers for handling, storage, and disposal of radioactive waste that are superior in performance, size and cost, while providing acceptable structural strength, shielding effectiveness, and carrying capacity.
In light of the shortcomings associated with existing dry storage containers and the need for long term management of existing inventories of SNF, the DOE began to examine alternative means for the transportation, storage and disposal of such waste. As a result of its investigation, the DOE recommended that the transport and emplacement of commercial spent fuel into a DOE waste repository be accomplished using a class of containers known as the Multi-Purpose Cask (MPC) and Multi-Purpose Unit (MPU). MPC/MPU containers are intended to perform the three functions of storage, transport, and disposal by direct emplacement into a waste repository. The MPC is a thin-shelled container, without shielding, which, once filled, is not intended to be opened. Proposed MPC/MPU designs use metal canisters requiring massive fabrication techniques. As a result, the estimated costs are three to six times greater than that of concrete cask designs. Furthermore, the MPC containers hold approximately 12% less SNF than that of concrete storage casks. Finally, since the MPC casks do not include shielding, these casks must be outfitted with overpacks consisting of thick-walled steel and, typically, a separate, neutron-absorbing material to provide shielding.
Meanwhile, the DOE was investigating management options and alternative uses for large quantities of depleted uranium hexafluoride (xe2x80x9cDUF6xe2x80x9d) stored at gas diffusion plants. Among the various disposal options considered by the DOE was conversion of the uranium hexafluoride to uranium metal, which could be machined for use as a radiation shielding material. However, the high costs of uranium metal production (around $10/kg), combined with the handling, machining, and environmental costs associated with the use of uranium metal have historically limited its use to only a few small applications. In connection with the design of the MPC and MPU, for example, the DOE proposed that depleted uranium metal be used as an axial shield plug in the MPC and as a gamma shielding material for the MPU during transport.
Other applications of depleted uranium metal in the fabrication of storage containers includes a container made from a composite containing a fibrous mat of interwoven metallic fibers encased within a concrete-based mixture that can include depleted uranium metal. Another proposed application includes a depleted uranium metal core for absorbing gamma rays and a bismuth coating for preventing chemical corrosion and absorbing gamma rays. Alternatively, a sheet of gadolinium may be positioned between the uranium metal core and the bismuth coating for absorbing neutrons. The containers can be formed by casting bismuth around a pre-formed uranium metal container having a gadolinium sheeting, and allowing the bismuth to cool.
Still another proposed application incorporates a depleted uranium metal wire wound on the inner shell of a cask to create a radiation shield. And yet another proposed application utilizes a composite radiation shield made up of rods of depleted uranium metal. The spaces between the rods contain smaller rods and are backfilled with lead or other high-density material. Still other designs utilize pipes of depleted uranium metal, tungsten, or other dense metal, encapsulating polyethylene cores, dispersed in rows of concentric bore holes around the periphery of the cask body. None of these existing designs, however, provides a simple, low-cost, low-weight radiation shielding system for transportation, storage, and disposal of radioactive waste.
Uranium compounds have also been proposed for use as shielding materials. For example, some investigators have proposed that depleted uranium dioxide (DUO2) pellets be mixed with a cement binder to form a material known as DUCRETE, which could be used as a shielding material in dry storage containers. The DUO2 pellets replace the gravel aggregate normally used in concrete. Due to the increased density of DUO2, however, the thickness of the shielding layer can be reduced. Thus, a storage container made from DUCRETE will have a greatly reduced weight and diameter compared to conventional concrete casks. In a typical cask, for example, the outer shell thickness can be reduced from approximately 2.5 feet for concrete to approximately one foot with DUCRETE. As a result, the cask diameter is reduced by approximately two-thirds, and the weight is reduced from approximately 123 tonnes to approximately 91 tonnes.
Despite these improvements in size and weight, however, DUCRETE casks systems suffer from disadvantages similar to those experienced with concrete casks. In particular, since DUCRETE has a low thermal conductivity and low temperature limit, DUCRETE casks must also incorporate labrynthine ventilation gaps. Furthermore, it is not expected that DUCRETE will be able to retain the uranium dioxide pellets in its cement matrix for a long period of time due to its high porosity of concrete and to the likelihood of water-cement-uranium dioxide reactions at warm temperatures (90-300xc2x0 C.). DUCRETE may also be incompatible with expected repository requirements. Hence, the use of DUCRETE in significant quantities for SNF disposal is questionable.
Nuclear fuel manufacturing plants produce small particles of uranium dioxide and uranium carbide by powdered metallurgical processes. These processes generally involve production of a powder of the proper particle size and range, which is then pressed into pellets, sintered, and ground to size. Even though powdered processes have shown success, their capacity is limited due to mechanical complexity, particle size, reactivity, and mass transfer limitations. In practice, line capacities are limited to approximately 100 tonnes/year, and maximum plant sizes to around 1,000 tonnes/year.
It has been proposed that aqueous processes be used to generate uranium dioxide and uranium carbide. Work on aqueous processes, and in particular on aqueous gelation processes, began in the late 1960""s. By the mid-1970""s pilot-scale facilities for production of uranium oxide and uranium carbide had been constructed. Experimental and pilot plant studies focused primarily on the use of uranyl nitrate solutions. For gelation, these uranyl nitrate solutions were dispersed using single nozzles into columns of chlorinated solvents such as trichloroethylene (TCE) and perchloroethylene. The resulting microspheres were then processed using multiple washing operations with water and ammonium hydroxide. The resulting microspheres, typically 0.03 mm to 2 mm in diameter, were incorporated into cylindrical pellets. Unfortunately, these aqueous processes had small throughputs and the processing was manually intensive. Thus, for planned capacities greater than 100 tonnes/yr, these processes were generally inadequate.
It is anticipated that the demand for shielding materials in accordance with the present invention will require the production of 5,000 to 30,000 tonnes/year of uranium dioxide and/or uranium carbide. Thus, there is a need for improved process capable of producing greater than 100 tonnes/year, and preferably 5,000-30,000 tonnes/year, of uranium dioxide and uranium carbide in reasonably-sized plants with inexpensive equipment. There is a further need for a process for producing microspheres of uranium dioxide and uranium carbide over a wide size range (30-1.200 microns). There is also a need for an improved gelation process for production of uranium dioxide and uranium carbide directly from uranium hexafluoride. Finally, there is a need for an improved gelation process that avoids the necessity of converting uranium hexafluoride to uranyl nitrate in order accomplish gelation. The present invention addresses these and other needs.
Briefly, and in general terms, the present invention resides in an improved radiation shielding material and storage systems for radioactive materials incorporating the same. The shielding material is preferably formed from a PYRolytic Uranium Compound (xe2x80x9cPYRUCxe2x80x9d) and provides improved radiation shielding in comparison with other shielding materials. In accordance with the invention, the shielding material can be used to form containment systems, container vessels, shielding structures, and containment storage areas, all of which can be used to house radioactive waste. The preferred embodiment of the shielding system is in the form of a container for storage, transportation, and disposal of radioactive waste.
The precursor for the PYRUC shielding material is preferably a mixture of a uranium compound and a binding material. In the preferred embodiment, the uranium compound is depleted uranium dioxide (DUO2) or depleted uranium carbide (DUC or DUC2). The uranium compound is preferably in the form of small particles, and more preferably in the form of pellets or microspheres, which can be coated or uncoated. The present invention incorporates a number of improvements over prior art methods for producing uranium dioxide and uranium carbide microspheres, whereby 5,000-30,000 tonnes/year of these microspheres can be produced in reasonably-sized plants and with inexpensive equipment. The improved gelation process of the present invention permits the use of oil in the gel forming column, deliberate carryover of oils to the sintering steps for supplying carbon and hydrogen, use of nitrogen as the sintering carrier gas, and use of peroxide for gelation of both uranium oxides and carbides.
In some cases, the precursor material can simply be cured to form a radiation shielding material. However, in preferred embodiments, the particles are immersed in a matrix of a binding material, so that the binding material fills the interstitial spaces and also provides additional neutron shielding. In accordance with the present invention, the binder is advantageously comprised of (1) a carbonaceous material (such as pitch); (2) a high-temperature resin (such as a polyimide); (3) a metal (such as aluminum powder); and/or (4) a metal-oxide (such as alumina). In addition, materials such as hydrogen, boron, gadolinium, hafnium, erbium, and/or indium in their non-radioactive isotopes, can be added in the mixture in the appropriate chemical form (usually the oxide) to provide additional neutron shielding effectiveness. The shielding materials are formed by applying sufficient heat to the mixture to cause a pyrolytic reaction that forms a solid material.
The present invention also resides in an method for manufacturing storage containers utilizing PYRUC shielding materials. In accordance with the invention, the precursor mixture can be poured or extruded into the container and then pyrolyized to form a solid shield. In a particularly preferred embodiment, the precursor starting materials are poured or extruded into a space formed by the inner and outer wall of a container and then pyrolized. The manufacturing process provides maximum flexibility in designing shielding shapes. The walls of the container provide the shape, structural support, and missile and drop protection, and also function as the secondary confinement barrier for the depleted uranium. The use of PYRUC simplifies shield manufacture and avoids the massive metal forging and machining activities associated with metal casks.
PYRUC shielding materials in accordance with the present invention offer superior gamma and neutron radiation shielding with the desirable thermal properties of metal at a much lower thickness, weight, and life-cycle cost than conventional materials. Furthermore, the PYRUC shielding materials can be optimized for specific circumstances and source terms. The use of depleted uranium reduces the assay (enrichment) level of the overall package, which provides for criticality mitigation. Furthermore, since PYRUC shielding materials have high thermal conductivities, the need for labyrinthine air passages and daily inspections is avoided. Similarly, PYRUC materials have higher thermal conductivities and temperature limits than concrete or DUCRETE and, thus, do not limit the design. In particular, the thermal conductivities of PYRUC materials exceed DUCRETE values by 25-100%. The temperature limits of carbonaceous PYRUC materials exceed 1000xc2x0 C. and PYRUC materials using other binders have temperature limits above 300xc2x0 C. Moreover, the high thermal conductivity and the high material temperature limit of PYRUC eliminate the need for a separate, inner canister for containing SNF. As a result, the PYRUC shielding materials can be used in SNF containers with direct contact between the shield""s inner annulus and the basket containing the SNF, which further reduces size and weight.
It is believed that PYRUC-shielded SNF containers will cost about $600,000 to $700,000 each, with the PYRUC component accounting for about $200,000 of the cost. The PYRUC container, although having an initial capital cost slightly greater than the concrete cask, is expected to be significantly less expensive than the metal cask while having similar advantages. Lower life-cycle costs are also expected for the PYRUC container as compared with either concrete or DUCRETE containers, since PYRUC""s superior heat transfer properties will preclude the need for frequent inspection and subsequent maintenance activities. Thus, PYRUC containers should be cost-competitive with traditional containers.
PYRUC is also environmentally desirable because it utilizes a waste product from the nuclear industry (depleted uranium) and, in one form, a waste product from the petrochemical industry (carbonaceous binder material) and converts them to environmentally stable forms. The PYRUC shielding material is also environmentally desirable because it is both microencapsulated and macroencapsulated, and has enhanced leach resistance. As a result, the material is potentially stable for geologic time periods. Thus, by virtue of its composition and expected behavior in a disposal environment, PYRUC is an environmentally friendly material.
Thus, the present invention satisfies the need for a shielding material having combined shielding performance, high temperature resistance, high thermal conductivity, and environmentally desirable characteristics, and for smaller, lighter containers for storage, transportation, and disposal of radioactive materials. While the primary applications for PYRUC are containers for SNF and HLW storage, transport, and disposal, PYRUC shielding materials can also be utilized in radiopharmaceutical containers, ion exchange resins, reactor cavity shielding and activated materials (i.e., made radioactive by neutron absorption) among others.
The present invention will be more clearly understood from a reading of the following detailed description in conjunction with the accompanying figures and tables.